In general, a boiling water reactor (BWR) adopts a primary reactor containment vessel of a pressure suppression type, and has hence an advantage of being able to reduce the volume of the primary reactor containment vessel. In the case of the boiling water reactor, the atmosphere in the primary reactor containment vessel is inerted with nitrogen gas to make the oxygen concentration lower than a concentration of oxygen contained in a normal air atmosphere. Therefore, if core fuel rises in temperature at the time of an accident and reacts with coolant to generate hydrogen, there is no risk of occurrence of detonation or deflagration in the primary reactor containment vessel. Further, the boiling water reactor adopts a small primary reactor containment vessel, and the outside of the primary reactor containment vessel is completely doubly housed in a reactor building. Thus, even in the case of a design basis accident, double confinement of radioactive materials can be ensured.
The safety of the boiling water reactor is extremely high, and there has been no outbreak of a design basis accident or an accident in which the integrity of the core fuel is impaired to the slightest degree at all. Further, according to the probabilistic safety assessment (PSA), it has been confirmed that the reactor core damage frequency of the boiling water reactor is as sufficiently low as about 10−5/reactor-year (r-y) to 10−6/reactor-year for old models and as further low as 10−7/reactor-year to 10−8/reactor-year. Even so, in the design of a next-generation reactor, a policy is adopted in which the design is made so as to ensure tolerance on the assumption that a severe accident could occur.
When an occurrence of a severe accident is assumed in the design of a next-generation reactor, there may be case where a large amount of hydrogen is generated upon occurrence of the severe accident due to metal-water reaction between high-temperature reactor fuel and coolant, and the pressure in the primary reactor containment vessel rises exceeding a design pressure limit. For example, in the case of an advanced boiling water reactor (ABWR), the design pressure of the primary reactor containment vessel is 310 kPa (45 psig); however, when a large amount of hydrogen is generated upon occurrence of a severe accident, the pressure in the primary reactor containment vessel may reach about twice the design pressure. In this regard, although it has been confirmed that the primary reactor containment vessel of the advanced boiling water reactor can endure a pressure three times the design pressure, a state where the pressure in the primary reactor containment vessel continues exceeding the design pressure for a long period of time upon occurrence of the severe accident is unfavorable in terms of safety. When the pressure in the primary reactor containment vessel exceeds the design pressure, radioactive gas inside the primary reactor containment vessel may leak at a leak rate higher than a design leak rate.
In the case of a design basis accident, the pressure in the primary reactor containment vessel is limited to a design pressure or less, and the leak rate of the radioactive gas inside the primary reactor containment vessel is limited to the design leak rate or less. In this case, the pressure in a reactor building hardly rises and, thus, the reactor building is not damaged at all. Further, in the case of the design basis accident, an active exhaust fan and a filter are used to discharge the gas in the reactor building through an exhaust pipe, thereby maintaining the pressure in the reactor building at a negative pressure relative to the atmospheric pressure. As a result, the reactor building has a double barrier against diffusion of radioactive materials to the atmosphere. However, upon occurrence of the severe accident, there is a possibility that the active exhaust fan fails due to loss of power or the like, which may cause a loss of the double confinement function of the reactor building.
In order to limit a pressure rise in the primary reactor containment vessel upon occurrence of the severe accident, it is effective to vent the gas in the primary reactor containment vessel to the reactor building serving as a secondary reactor containment vessel. However, the reactor building has a low design pressure, and the gas therein is normal air, so that if high-pressure gas in the primary reactor containment vessel containing a large amount of hydrogen generated upon occurrence of the severe accident is discharged to the reactor building, detonation may occur in the reactor building to damage the reactor building, resulting in uncontrolled release of radioactive gas to the atmosphere. This situation resembles the situation of the severe accident in Chernobyl nuclear plant. Thus, this method is not technically feasible, although having been proposed, and there has been no actual boiling water reactor that adopts such a dangerous method until now. Incidentally, the design pressure of the reactor building is as low as 13.8 kPa (2 psig or 0.14 kg/cm2). However, this value does not mean that a conventional reactor building is fragile but is intentionally set as the pressure at which a blow-out panel, which is installed for safety purpose, is to be opened.
As a method that is technically feasible, there is known one in which the gas in the primary reactor containment vessel upon occurrence of the severe accident is released to the environment through a filter. However, radioactive noble gas and organic iodine are released to the environment, without being eliminated, so that radiation exposure to the public occurs. Further, in order to reduce the radiation exposure to the public to the minimum, it is necessary to previously evacuate every people living within a certain distance.
Recently, in an ESBWR, an excellent method is adopted in which steam to be released to the primary reactor containment vessel upon occurrence of an accident is condensed by a passive containment cooling system (PCCS) to suppress a pressure rise due to the decay heat of the primary reactor containment vessel. However, a large amount of hydrogen released upon occurrence of a severe accident is not condensed by the PCCS, and the volume of the primary reactor containment vessel of the ESBWR is small, so that the pressure in the primary reactor containment vessel upon occurrence of the severe accident is maintained at twice or three times the design pressure after all. That is, even when the PCCS is installed in the conventional reactor containment vessel having a small volume, the pressure in the reactor containment vessel upon occurrence of the severe accident cannot be limited to the design pressure or less. Incidentally, the design pressure of the primary reactor containment vessel of the ESBWR is also 310 kPa (45 psig).
Under the above circumstances, recently, there has been disclosed an innovative reactor containment vessel having a configuration in which an upper space including an operation floor is used as a pressure-resistant upper secondary reactor containment vessel and high-pressure gas in the primary reactor containment vessel is vented to the secondary reactor containment vessel (See Jpn. Pat. Appln. Laid-Open Publication No. 2007-10457, the entire content of which is incorporated herein by reference.). The pressure resistance in this case means that the upper secondary reactor containment vessel is designed to have substantially the same design pressure as that of the primary reactor containment vessel. However, in this reactor containment vessel, there was a necessity to make the atmosphere in the space including an operation floor lower in oxygen concentration than normal air atmosphere in order to prevent occurrence of hydrogen detonation. In this circumstance, operators cannot enter the space including an operation floor during normal operation of the plant for, e.g., fuel transportation. If the atmosphere of the space including an operation floor is air, a large amount of hydrogen generated upon occurrence of the severe accident needs to be burned or recombined by an igniter or a recombiner, generating a large amount of heat generation to raise the temperature of the reactor containment vessel to 200° C. or more, which may cause overtemperature failure.
In the case of a next-generation reactor, the designed lifetime of a plant is as long as 60 years to 80 years and thus there is a possibility that the number of residents in the vicinity of the plant increase after the construction of the plant. The increase in the number of the peripheral residents makes it difficult to fulfill complete evacuation of all the residents, which is required to be carried out as an emergency planning upon occurrence of the severe accident. When the number of the peripheral residents is limited, it becomes difficult to construct a nuclear plant in the vicinity of a heavily-populated large city. Even if being constructed in the vicinity of a densely populated large city, the safety of a next-generation reactor must be fully guaranteed by the inherent design safety of the plant.
Further, the next-generation reactor is likely to be constructed in any location all over the world and is therefore at risk of encountering a natural disaster such as a giant earthquake, a large tsunami, and a mega hurricane. When a severe accident occurs due to such severe natural disasters, it is difficult to fulfill complete evacuation of all the residents in the vicinity of the plant. We must understand that the next-generation reactors of all over the world are at risk of encountering very sever natural disasters, such as a mega cyclone, a giant earthquake, and a large tsunami. If a severe natural disaster such as a great cyclone occurs followed by a station blackout (SBO), there is a possibility that recovery work cannot be carried out for a long time period. Thus, it is necessary to provide a next-generation reactor safe enough to prevent the consequence of the severe accident due to such a prolonged station blackout from reaching over the offsite and to eliminate the need for evacuation of the peripheral residents.
As described above, it is desirable to provide the next-generation reactor safe enough to eliminate the need for evacuation of the peripheral residents if the severe accident occurs due to a devastating natural disaster. To this end, it is necessary to conduct design so as to maintain the pressure in the reactor containment vessel upon occurrence of the severe accident at a design pressure or less, to prevent the integrity of the reactor containment vessel from being impaired due to occurrence of detonation or deflagration of a large amount of hydrogen, and to prevent radioactive noble gas or organic iodine from being released to the environment through a filter vent. Further, it is desirable to provide a configuration in which leakage of radioactive materials from the reactor containment vessel can be passively prevented without use of an external power source. Further, at the normal operation time of the plant, it is desirable that operators can conduct their work on the operation floor.